This invention relates to a process and apparatus for solidification of radioactive waste occurring in a nuclear power station, and more particularly to a process and apparatus for its solidification utilizing a hydraulic solidifier.
The amounts of radioactive waste occurring in nuclear power stations and related facilities have been increasing year by year, and a need for the volume reduction of such radioactive waste has consequently been increasing in order to secure a storage space within the facilities.
Methods which have so far been examined for the volume reduction of radioactive waste includes the following. Concentrated liquid waste obtained by concentrating the liquid waste formed in the regeneration of spent ion-exchange resin and the slurry of powdery ion-exchange resin which occur in large amounts in a nuclear power station are dried into powder. Thus the liquid waste is freed of water which accounts for the major or part of its volume and, if necessary, the powder is further pelletized and solidified collectively by packing in a solidification vessel. (U.S. Pat. No. 4,299,271).
However, such a method is still defective in that the liquid waste cannot necessarily be converted into a stable solid when using a hydraulic solidifier such as cement or alkali silicates (e.g. water glass).
The concentrated liquid waste occurring in a boiling water reactor (BWR) nuclear power station is composed chiefly of a sodium salt, i.e. sodium sulfate (Na.sub.2 SO.sub.4). In a pressurized water reactor (PWR) nuclear power station, on the other hand, the concentrated liquid waste is composed chiefly of a sodium salt, i.e. sodium borate (Na.sub.2 B.sub.4 O.sub.7). These sodium salts are both water-soluble.
In case the concentrated liquid waste occurring in a BWR nuclear power station is dried, powdered or, if necessary, further pelletized, and then solidified with a hydraulic solidifier, sodium sulfate which is its main ingredient will absorb free water contained in the solidifier paste and water formed by the solidification reaction, and thereby form a swollen hydrate, Na.sub.2 SO.sub.4.10H.sub.2 O, to cause cracking in the solidified body. In addition, in case of a cement solidifier, sodium sulfate will react with calcium hydroxide which is formed when cement is hydrated, and thereby form gypsum, which will prevent the cement from hardening too rapidly but will, on the other hand, accelerate the formation of ettringite (3CaO.Al.sub.2 O.sub.3.3CaSO.sub.4.32H.sub.2 O) to cause the solidified body to be swollen or broken.
In case the concentrated liquid waste occurring in a PWR nuclear power station is solidified, sodium borate, which is the main ingredient, will likewise cause the solidified body to lower its strength. It will form a hydrage, Na.sub.2 B.sub.4 O.sub.7.10H.sub.2 O, to generate heat. In the case of a cement solidifier, it will inhibit the formation of a hydrate of calcium silicate (3CaO.2SiO.sub.2.3H.sub.2 O) and of a hydrate of calcium aluminate (3CaO.Al.sub.2 O.sub.3.6H.sub.2 O) by the hydration of cement.
Since, in either of these cases, the powdered or pelletized waste mainly comprises the water-soluble sodium salts, the solidified body suffers from degradation of its structure, reduction in the leaching rate, and lowering in the strength and specific gravity owing to exudation during a prolonged storage.
In the solidification procedure, furthermore, sodium borate reacts with the hydraulic solidifier very promptly, and the solidification proceeds so rapidly as to disturb the smooth pouring of the solidification mixture. To prevent this, the content of the liquid waste in the solidification mixture will have to be limited to at most 30 wt % and the volume reduction ratio be correspondingly lowered.